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An important area of research required for fusion reactor design is the study of materials under high energy neutron irradiation. Deuterium-Tritium (D-T) reactions release 14.1 MeV neutrons and material studies of such high energy neutrons focusing on transmutation and activation are paramount for fusion tokamak devices such as ITER and DEMO. In order to understand neutron damage and transmutation-induced radioactivity in fusion regime energies, a series of experimental campaigns were performed at the ASP facility based at Aldermaston in the UK, which uses a deuteron accelerator to bombard a tritiumloaded target and generate 14 MeV-neutron emission rates of up to 2.5 × 10 11 s −1 . In this work, a holistic treatment of the 11,000 gamma spectra (time series data) collected over five experimental campaigns is applied to identify radioisotopes and validate nuclear data and the inventory code, FISPACT-II. Whilst previous analysis has examined single spectra and foil irradiation’s using traditional, human-driven methods, this work applies novel methods using Artificial Neural Networks (ANN) and classification algorithms to allow a fully automated approach. Using such methods we show good broad agreement with FISPACT-II inventory simulations, and an overview of results are given as C/E values.

For either nuclear fusion or generation IV fission reactors to be viable as a commercial energy source the decommissioning and waste disposal solutions must be considered during the design. A multi-step simulation process combining Monte Carlo Neutron Transport simulations with inventory simulations have been performed to estimate the activation of steels in key reactor components of the European Sodium-cooled fast Reactor (ESFR). Waste classifications based on UK waste disposal regulations have been applied to the key components to estimate the expected masses of low level and intermediate level waste. The use of reduced activation steels, EUROFER and F82H, in reactor components external to the core results in a factor of 10 reduction in the percentage of waste classified as Intermediate Level Waste (ILW). Waste estimates are compared to existing waste estimates for the European Demonstration fusion reactor (DEMO). The ESFR has a lower percentage of ILW per total reactor mass due to irradiated steels compared to DEMO. However, there is no Higher Activity Waste (HAW) associated with DEMO, compared with arisings from the ESFR spent fission fuel.

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR.

With ongoing advances in fusion and advanced fission reactors, quantifying irradiation effects in materials is critical. Transmutation-induced helium in cladding and structural materials can drive swelling and embrittlement, thereby reducing these components’ lifespans. Yet most studies ignore the considerable uncertainties in predicting helium generation rates. In this work, we created a code wrapper, F-SCATTER, that automatically performs simulations in FISPACT-II. We used this tool to investigate potential variance in helium generation rate, or He/dpa, calculations based on deviations in alloy composition, irradiating neutron flux spectrum, computational methodology, and nuclear data sources. We used 12 wt% Cr HT9 steel as the reference case and observed a 6.5%–98.3% He/dpa spread based on compositional variation within a single chemical specification, a 1.8%–11.5% He/dpa variation upon the incorporation of a 15% artificial uncertainty in flux at each energy, and a He/dpa difference as high as 231% when using ENDF/B-VIII.0 versus TENDL-2021 data libraries. Similar results were found for other prominent iron-based alloys, including Grade 91, castable nano-structured alloy, and 316H—where additional variations exist based on reactor type (e.g. thermal, fast, or fusion) and alloying elements such as carbon, nitrogen, and nickel. Based on the simulated results, we conclude that a significant part of the heat-to-heat variability in swelling responses of Fe-based alloys can be driven by impurity content in alloy compositions, and, therefore, chemical control should be a key element in supply chain design for advanced nuclear energy systems. Furthermore, we provide critical recommendations on best practices for evaluating and reporting helium production and lattice damage rates when computing predictions with multiphysics programs such as FISPACT-II.

Various substances in the target room of 70-MeV proton beam irradiation facility must be managed as radioactive waste owing to their activation during operations. In radioactive waste management, it is important to calculate the storage period. Consequently, radionuclides of activated substances and radioactivity concentration must be investigated. This study investigated the radionuclides and radioactivity values of activated aluminum, concrete, stainless steel, and each concrete by depth using computational simulations (Monte Carlo N-Particle, FISPACT-II) and measurements. Most of the radionuclides and radioactivity values of the substances obtained by the two methods exceeded the clearance level. For nuclides exceeding the clearance level, the ratio of the FISPACT-II to measurement results yielded an average of 1.21. With increase in the depth of the concrete, the difference between the two results increased. For all substances, the period of satisfying the clearance criterion was more than 10 years. This long period is attributed to the effects of long half-life nuclides 22 Na and 54 Mn. Thus, when calculating the storage period for radioactive waste disposal, the method resulting in higher activity values for 22 Na and 54 Mn should be applied.

Nuclear activation is the process of production of radionuclides by irradiation. This phenomenon concerns particle accelerators used in various fields, from medical applications to industrial ones, both during operation and at the decommissioning phase. For more than three decades, the possibility of using cyclotrons for nuclear power generation and nuclear waste reduction has also been discussed, i.e. in the case of Accelerator-Driven Systems [1]. The radioprotection and dismantling issues of accelerator facilities, that have been raised recently, is even more potent for such installations. In our study, we are particularly interested in the activation due to secondary neutrons produced by (x,n) reactions, mostly (p,n) occurring in the accelerator’s components. This work focuses on the study of the radioactivity induced in various materials (V, Sc, Tb, W, Ta) irradiated by fast and thermal neutrons, in two different scenarios: through direct irradiation -with an AmBe sourceand around an operating cyclotron at the CYRCé facility (Strasbourg). A broad Monte Carlo study including FLUKA, GEANT4, PHITS and MCNP simulation has been performed, with and without a FISPACT-II coupling, to estimate the reaction rates and to trace the induced radioactivity in samples of known composition. The results of the simulations are compared with the values extracted in two dedicated experimental campaigns in which activated samples underwent high resolution gamma-ray spectrometry.

Retrospective characterization of a (30 MeV D, Be) neutron source was performed employing multi-foil activation and STAYSL-PNNL. Experimental reaction rates were calculated from gamma spectroscopy measurements of irradiated foils and MCNP provided the guess spectrum. Adjusted spectra were evaluated through activation calculations for a stainless-steel target using FISPACT-II. Adjusted spectra showed limited dependence on the dosimetry reactions and provided minor improvements in activation calculations. Omitting reflected neutrons in the guess spectrum generated poor activation results and the limited number of dosimetry reactions introduced doubt in the adjusted spectra. A dedicated neutron spectrometry experiment and a more detailed simulation is required.